A full-flow-range drift-flux model for adiabatic and boiling two-phase flows in vertical narrow rectangular flow channels Xiuzhong Shen, Toshihiro Yamamoto, Ken Nakajima, Takashi Hibiki International Journal of Thermal Sciences, 2024 This study has reviewed the past research for adiabatic and boiling GLTP (gas-liquid two-phase) flows in VNR (vertical narrow rectangular) flow channels. 1758 void fraction experimental data have been collected from 13 available databases of various researchers. The underlying distribution parameter and drift velocity features have been obtained by analyzing the experimentally-measured data in the classical Zuber-Findlay plots. Existing seven DF (drift-flux) correlations for the VNR flow channels have been examined with the obtained distribution parameter and drift velocity features from Zuber-Findlay plots and the collected 1758 void fraction data and their predictions are found to be unsatisfactory. So, this study has analyzed the dependencies of the distribution parameter and drift velocity on two-phase density ratio, flow channel separation and width, and local flow conditions, and developed new flow-regime-independent full-flow-range distribution parameter and the drift velocity correlations for the VNR flow channels. These newly-developed correlations have been examined with the obtained distribution parameter and drift velocity features from Zuber-Findlay plots and their agreement has been confirmed. The DF model closed by the newly-developed distribution parameter and the drift velocity correlations has also been verified by the collected 1758 void fraction data, and their relative error is 13.1 %.
Experimental investigation on local flow structures of upward cap-bubbly flows in a vertical large-size square channel Haomin Sun, Tomoaki Kunugi, Takehiko Yokomine, Xiuzhong Shen, Takashi Hibiki Experimental Thermal and Fluid Science, 2024 Gas–liquid two-phase flows are typical in engineering systems involving heat and mass transfers. The cross-sectional geometries and sizes of these flow channels vary, exerting a notable impact on the thermofluid behavior. To develop thermofluid models, it is imperative to obtain local measurements of two-phase flows and gain insights into their characteristics from these measurements. Numerous experimental studies have investigated the local flow characteristics in circular pipes and rectangular channels; however, few studies on square channels have been conducted, particularly for flow regimes beyond bubbly flows, such as cap-bubbly flows. Given the importance of two-phase flows in large square channels for advanced nuclear reactors, such as the economic simplified boiling water reactor (ESBWR), we experimented with upward cap-bubbly flows in a large square channel. Detailed cross-sectional distributions of the void fractions, axial gas velocities, and interfacial area concentrations for two bubble-size groups were obtained at three axial locations using local measurements of a four-sensor optical probe. Based on the database, the cap-bubbly flow characteristics were understood, including flow development in a large square channel. In addition, the existing drift-flux correlations for large circular pipes predicted the measured void fraction with an accuracy of approximately ± 10 %, whereas the existing interfacial area concentration correlations reasonably correlated with the measured interfacial area concentration with an accuracy of approximately ± 30 %. Furthermore, the database was used to calculate the covariances of the void fractions. The correlations for large circular pipes significantly underestimated the void fraction covariance of the larger bubble group, probably because of the peculiar void fraction distribution of the group in a large square channel. The other covariances were well predicted.
Axial flow characteristics of bubbly flow in a vertical large-diameter square duct 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics Nureth 2017, 2017
Two-group interfacial area concentration correlations of two-phase flows in large diameter pipes International Conference on Nuclear Engineering Proceedings ICONE, 2015
Evaluation Of interfacial area transport equation in coupled two-fluid model computation International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015 Nureth 2015, 2015
Implementation of two-group interfacial are a transport in a one-dimensional computational environment Embedded Topical Meeting on Advances in Thermal Hydraulics Ath 2014 Held at the American Nuclear Society 2014 Annual Meeting, 2014
Shape measurement of bubble in a liquid metal Y. Saito, X. Shen, K. Mishima, M. Matsubayashi Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment, 2009
